عنوان

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems -- Water Reactors :

3030046397

9783030046392

3030046389

9783030046385

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems -- Water Reactors :

[Book]

Portland, Oregon, August 13-17, 2017 /

John H Jackson, Denise Paraventi, Michael Wright, editors.

Cham, Switzerland :

Springer,

c2018.

1 online resource.

The minerals, metals & materials series,

2367-1181

International conference proceedings.

Includes bibliographical references and indexes.

Part 1. PWR Nickel SCC -- SCC.- Scoring Process for Evaluating Laboratory PWSCC Crack Growth Rate Data of Thick-wall Alloy 690 Wrought Material and Alloy 52, 152, and Variant Weld Material.- Applicability of Alloy 690/52/152 Crack Growth Testing Conditions to Plant Components.- SCC of Alloy 152/52 Welds Defects, Repairs and Dilution Zones in PWR Water.- NRC Perspectives on Primary Water Stress Corrosion Cracking of High-chromium, Nickel-based Alloys.- Stress Corrosion Cracking of 52/152 Weldments near Dissimilar Metal Weld Interfaces.- Composite Material Stress Corrosion Crack Arrest Testing in Hydrogen Deaerated Water.- Investigation of Hydrogen Behavior in Relation to the PWSCC Mechanism in Alloy TT690.- Part 2. PWR Nickel SCC -- Initiation.- Crack Initiation of Alloy 600 in PWR Water.- SCC Initiation Behavior of Alloy 182 in PWR Primary Water.- Multiple Cracks Interactions in Stress Corrosion Cracking: In-situ Observation by Digital Image Correlation and Phase Field Modelling.- Stress Corrosion Cracking Initiation of Alloy 82 in Hydrogenated Steam.- Application of Ultra-high Pressure Cavitation Peening on Reactor Vessel Head Penetration, BMN and Primary Nozzles.- The Effect of Surface Condition on Primary Water Stress Corrosion Cracking Initiation of Alloy 600.- Microstructural Effects on SCC Initiation in Simulated PWR Primary Water for Cold-worked Alloy 600.- Part 3. PWR Nickel SCC -- Aging Effects.- A Kinetic Study of Order-disorder Transition in Ni-Cr Based Alloys.- The Role of Stoichiometry on Ordering Phase Transformations in Ni-Cr Alloys for Nuclear Applications.- The Effect of Hardening via Long Range Order on the SCC and LTCP Susceptibility of a Nickel-30Chromium Binary Alloy.- PWSCC Initiation of Alloy 600: Effect of Long-term Thermal Aging and Triaxial Stress.- Stress Corrosion Cracking Behavior of Alloy 718 Subjected to Various Thermal Mechanical Treatments in Primary Water.- Time- and Fluence-to-fracture Studies of Alloy 718 in Reactor.- Development of Short-range Order and Intergranular Ccarbide Precipitation in Alloy 690 TT upon Thermal Ageing.- Part. 4. PWR Nickel SCC -- Alloy 600 Mechanistic.- Diffusion Processes as a Possible Mechanism for Cr Depletion at SCC Crack Tip.- Role of Grain Boundary Cr5B3 Precipitates on Intergranular Attack in Alloy 600.- Advanced Characterization of Oxidation Processes and Grain Boundary Migration in Ni Alloys Exposed to 480 °C Hydrogenated Steam.- Exploring Nanoscale Precursor Reactions in Alloy 600 in H2/N2-H2O Vapor Using In Situ Analytical Transmission Electron Microscopy.- Electrochemical and Microstructural Characterization of Alloy 600 in Low Pressure H2origin: initial; background-clip: initial;">-Steam.- Effect of Dissolved Hydrogen on the Crack Growth Rate and Oxide Film Formation at the Crack Tip of Alloy 600 Exposed to Simulated PWR Primary Water.- A Mechanistic Study of the Effect of Temperature on Crack Propagation in Alloy 600 under PWR Primary Water Conditions.- Part 5. PWR Nickel SCC -- Alloy 690 Mechanistic.- Grain Boundary Damage Evolution and SCC Initiation of Cold-worked Alloy 690 in Simulated PWR Primary Water.- Effect of Cold Work and Grain Boundary Carbides on PWSCC Susceptibility of Alloy 690.- Relationship among Dislocation Density, Local Strain Distribution, and PWSCC Susceptibility of Alloy 690.- Morphology Evolution of Grain Boundary Carbides Precipitated near Triple Junctions in Highly Twinned Alloy 690.- A Mechanistic Study on the Stress Corrosion Crack Propagation for Heavily Cold Worked TT Alloy 690 in Simulated PWR Primary Water.- Microstructural Study on the Stress Corrosion Cracking of Alloy 690 in Simulated Pressurized Water Reactor Primary Environment.- Part 6. Effect of Strain Rate and High Temperature Water on Deformation Structure of VVER Neutron Irradiated Core Internals Steel.- Radiation-Induced Precipitates in a Self-Ion Irradiated Cold-Worked 316 Austenitic Stainless Steel Used for PWR Baffle-Bolts.- In Situ and Ex Situ Observations of the Influence of Twin Boundaries on Heavy Ion Irradiation Damage Effects in 316L Austenitic Stainless Steels.- In Situ Microtensile Testing for Ion Beam Irradiated Materials.- Development of High Irradiation Resistance and Corrosion Resistance Oxide Dispersion Strengthed Austenitic Stainless Steels.- Probing Damage Gradients in Ion-irradiated Tungsten Using Spherical Nanoindentation.- Part 7. Irradiation Damage -- Swelling.- Formation of He Bubbles by Repair-welding in Neutron-irradiated Stainless Steels Containing Surface Cold Worked Layer.- Predictions and Measurements of Helium and Hydrogen in PWR Structural Components Following Neutron Irradiation and Subsequent Charged Particle Bombardment.- Emulating Neutron-induced Void Swelling in Stainless Steels Using Ion Irradiation.- Carbon Contamination, Its Consequences and Its Mitigation in Ion-simulation of Neutron-induced Swelling of Structural Steels.- Void Swelling Screening Criteria for Stainless Steels in PWR Systems.- Theoretical Study of Swelling of Structural Materials in Light Water Reactors at High Fluencies.- Part 8. Irradiation Damage -- Nickel Based and Low Alloy.- High Resolution Transmission Electron Microscopy of Irradiation Damage in Inconel X-750.- In-situ SEM Push-to-pull Micro-tensile Testing of in Service Inconel X-750 Annulus Spacers.- Microstructural Characterization of Proton-irradiated 316 Stainless Steels by Transmission Electron Microscopy and Atom Probe Tomography.- Part 9. PWR Stainless Steel SCC and Fatigue -- SCC.- Microstructural Effects on Stress Corrosion Initiation in Austenitic Stainless Steel in PWR Environments.- Oxidation and SCC Initiation Studies of Type 304L SS in PWR Primary Water.- SCC Initiation in the Machined Austenitic Stainless Steel 316L in Simulated PWR Primary Water.- High-resolution Characterisation of Austenitic Stainless Steel in PWR Environments: Effect of Strain and Surface Finish on Crack Initiation and Propagation.- SCC of Austenitic Stainless Steels under Off-normal Water Chemistry and Surface Conditions Part I: Surface Conditions and Baseline Tests in Nominal PWR Primary Environment.- SCC of Austenitic Stainless Steels under Off-normal Water Chemistry and Surface Conditions Part II: Off Normal Chemistry -- Long Term Oxygen Conditions and Oxygen Transients.- The Effect of Microchemistry on the Crack Response of Lightly Cold Worked Dual Certified Type 304/304L Stainless Steel after Sensitizing Heat Treatment.- Part 10. PWR Stainless Steel SCC and Fatigue -- Fatigue.- The Effect of Load Ratio on the Fatigue Crack Growth Rate of Type 304 Stainless Steels in Air and High Temperature Water at 482\\176F.- Electrical Potential Drop Observations of Fatigue Crack Closure.- The Effect of Environment and Material Chemistry on Single-Effects Creep Testing of Austenitic Stainless Steels.- Corrosion Fatigue Behavior of Austenitic Stainless Steel in Pure D2O Environment.- Mechanistic Understanding of Environmentally Assisted Fatigue Crack Growth of Austenitic Stainless Steels in PWR Environments.- Study on Hold-Time Effects in Environmental Fatigue Lifetime of Low-alloy Steel and Austenitic Stainless Steel in Air and under Simulated PWR Primary Water Conditions.- Part 11. Special Topics I -- Materials.- Evaluation of Additively Manufactured Materials for Use as Nuclear Plant Components.- Hot Cell Tensile Testing of Neutron Irradiated Additively Manufactured Type 316L Stainless Steel.- Computational and Experimental Studies on Novel Materials for Fission Gas Capture.- Hydrogen Assisted Cracking Studies of a 12% Chromium Martensitic Stainless Steel -- Influence of Hardness, Stress and Environment.- Investigation of Flow Accelerated Corrosion Models to Predict the Corrosion Behavior of Coated Carbon Steels in Secondary Piping Systems.- Effect of High-Temperature Water Environment on the Fracture Behaviour of Low-alloy RPV Steels.- Corrosion Fatigue Testing of Low Alloy Steel in Water Environments with Low Levels of Oxygen and Varied Load Dwell Times.- U-1: Feasibility Study of the Internal Zr/ZrO2 Reference Electrodes in Supercritical Water Environments.- Part 12.

Special Topics II -- Processes.- Investigation Of Pitting Corrosion In Sensitized Modified High-Nitrogen 316LN Steel After Neutron Irradiation.- Quantifying Erosion-corrosion Impacts on Light Water Reactor Piping.- Effect of Molybdate Anion Addition on Repassivation of Corroding Crevice in Austenitic Stainless Steel.- Effect of pH on Hydrogen Pick-up and Corrosion in Zircaloy-4.- Oxidation Kinetics of Austenitic Stainless Steels as SCWR Fuel Cladding Candidate Materials in Supercritical Water.- A Recent Look at CANDU Feeder Cracking: High Resolution Transmission Electron Microscopy and Electron Energy Loss near Edge Structure (ELNES).- Part 13. Cables and Concrete Aging and Degradation -- Cables.- Simultaneous Thermal and Gamma Radiation Aging of Electrical Cable Polymers.- Principal Component Analysis (PCA) as a Statistical Tool for Identifying Key Indicators of Nuclear Power Plant Cable Insulation Degradation.- How Can Material Characterization Support Cable Aging Management?.- Aqueous Degradation in Harvested Medium Voltage Cables in Nuclear Power Plants.- Frequency Domain Reflectometry Modeling and Measurement for Nondestructive Evaluation of Nuclear Power Plant Cables.- Aging Mechanisms and Nondestructive Aging Indicator of Filled Cross-linked Polyethylene (XLPE) Exposed to Simultaneous Thermal and Gamma Radiation.- Successful Detection of Insulation Degradation in Cables by Frequency Domain Reflectometry.- Capacitive Nondestructive Evaluation of Aged Cross-Linked Polyethylene (XLPE) Cable Insulation Material.- C-2: Tracking of Nuclear Cable Insulation Polymer Structural Changes using the Gel Fraction and Uptake Factor Method.- C-4: Degradation of Silicone Rubber Analyzed by Instrumental Analyses and Dielectric Spectroscopy.- Part 14. Cables and Concrete Aging and Degradation -- Concrete.- Automated ......

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Springer Nature

com.springer.onix.9783030046392

9783030046385

Conference on environmental degradation of materials in nuclear power systems 2018

Nuclear power plants-- Corrosion, Congresses.

Nuclear power plants-- Materials-- Effect of radiation on, Congresses.

Water cooled reactors-- Corrosion, Congresses.

Nuclear power plants-- Corrosion.

Nuclear power plants-- Materials-- Effect of radiation on.

TECHNOLOGY & ENGINEERING / Mechanical.

Water cooled reactors-- Corrosion.

TEC-- 009070

Jackson, John H.

Paraventi, Denise

Wright, Michael

International Conference on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors(18th :2017 :, Portland, Or.)

Minerals, Metals and Materials Society.

20200823080753.0

pn

[Book]

Y